船用堆控制棒失控抽出事故时堆芯安全特性研究

采用船用堆三维动态安全分析仿真软件对发生控制棒失控抽出事故时堆芯安全特性进行了仿真分析,研究了反应堆分别处于高、低功率运行工况下1组或1束控制棒以不同的速率失控抽出时堆芯燃料芯块中心最高温度、最小烧毁比和冷却剂出口温度等参数的变化规律,并进行了比较,得出了一些有益的结论,对于考察反应堆安全状况和事故发生后反应堆操纵人员制定安全措施具有重要的指导意义。     

A linear equation of characteristic time profile of power in subcritical quasi-steady state

ABSTRACT

An equation of power in subcritical quasi-steady state has been derived based on one-point kinetics equations for the purpose of utilizing it for the development of timely reactivity estimation from complicated time profile of neutron count rate. It linearly relates power, P, to a new variable q, which is a function of time differential of the power.

It has been confirmed that the points (q, P) calculated by using one-point kinetics code, AGNES, are perfectly in a line described by the new equation and that the points (q, P) calculated from transient subcritical experiment data measured by using TRACY made a line with a slope indicated by the new equation.     

混合能谱超临界水冷堆反应性引入事故安全分析

本文研究了混合能谱超临界水冷堆(SCWR-M)在发生控制棒失控提升事故和弹棒事故这两类反应性引入事故后的反应堆系统响应。首先利用修改的可用于超临界条件下的系统程序RELAP5对混合能谱超临界水冷堆进行系统建模,并计算分析在功率运行工况下事故过程中功率、流量及包壳温度等重要参数的变化趋势,最后对反应性参数如控制棒价值、控制棒抽出速率和负反馈系数进行了参数效应分析。结果表明,在设计工况下混合能谱超临界水冷堆系统可有效地将衰变热导出堆芯,保证了燃料棒的完整性。另外,反应性参数对控制棒失控提升事故的安全性影响不大,但对弹棒事故的包壳峰值温度影响很大,过于保守的反应性参数估计会使安全裕量大为减小。     

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.     

Significant spatial dependence observed in inverse kinetics analysis for a loosely coupled-core system of the Kyoto University Critical Assembly

A loosely coupled-core system was constructed in the Kyoto University Critical Assembly to study the spatial effect observed in inverse kinetics analysis of control rod reactivity worth. In a rod drop experiment, the conventional inverse kinetic method resulted in a space- and time-dependent rod worth, which depended significantly on detector position and varied remarkably with the elapse of time. In another rod worth measurement, where a control rod was continuously inserted, the similar spatial dependence could be also observed. In this study, a modal expansion approach was proposed to reduce the above spatial dependence of the measured rod worth. Applying the present approach to inverse kinetics analysis, the troublesome dependence could be solved to obtain space-independent rod worth. This approach requires only the eigenfunctions of fundamental and higher modes for an unperturbed system but makes both static and transient calculations for various perturbed systems unnecessary.     

Experimental Study on Reactivity Worth for Absorber Material in High Conversion Light Water Reactor Using FCA-HCLWR Core Fueled with Enriched Uranium

Experimental study on reactivity worth for absorber material in HCLWR core has been carried out in a series of experiments using the Fast Critical Assembly (FCA) in Japan Atomic Energy Research Institute (JAERI). The central reactivity worth as well as the simulated control rod worth of B₄C with different ¹⁰B content and of Hf was measured in FCA-HCLWR core fueled with enriched uranium. Both reactivity worths of B₄C increase with ¹⁰B content. These increasing trends do not saturate to 90% enriched B₄C. The Hf has the smaller reactivity worth than the 20% B₄C. The experimental values are compared with the calculated ones which obtained from JENDL-2 data and the SRAC system. The calculation predicts well the dependence of reactivity worth on ¹⁰B content and underestimates the reactivity worth ratios of the Hf to the 20% B₄C.     

压水堆燃料元件制造设施安全相关事件的统计和分析

描述了我国压水堆燃料制造设施1987～2005年安全相关事件的统计和分析。结果表明,事件主要包括与安全相关的系统和重要设备故障、导致密封屏障失效或损坏事件、可能导致临界的事件和其他事件,它们占事件总数的68.9%。对事件原因和事件后果作了统计分析。最后,为减少事件的发生提出了一些建议。     

Nuclear Safety Analysis of a Molten Salt Breeder Reactor

An analysis is made of the nuclear safety of a Molten Salt Breeder Reactor (MSBR) using one point kinetics model, with particular attention to the fact that the MSBR, operating on ²³³U-Th with circulating fuel, has a very small effective delayed neutron fraction, which is conducive to a very high power increase being induced by reactivity addition, particularly when applied sharply.

The analysis proves that the transient phenomena of MSBR initiated by an abnormal reactivity addition do not exceed the bounds of structural safety, in virtue of the very favorable nuclear and thermal characteristics of the MSBR.     

截短型燃料组件堆内贮存临界安全分析

基于小型压水堆特有的截短型燃料组件,建立乏燃料贮存水池几何模型,分析正常贮存及事故工况下的临界安全。选取合理的保守假设,建立适当的计算模型,分别计算了一区和二区正常贮存工况、地震事故工况、组件跌落事故工况、新组件误插入事故工况的反应性。计算得到事故工况下有效增值因子最大值为0.932 83。小型压水堆乏燃料贮存水池临界安全分析中,正常工况及事故工况下计算结果均小于0.95。该设计模型可确保燃料堆内贮存区域处于次临界状态,且安全可控。     

Interpretation of Spatial Kinetics for Rod Drop Experiment in Loosely Coupled-Core Systems

Abstract

The coupled two-core reactor systems with various degrees of spatial coupling were constructed in the Kyoto University Critical Assembly (KUCA) to study the spatial kinetics observed in the control rod drop experiment. By applying the two-mode and the two-point kinetic models to the space-dependent rod worths measured on the basis of the one-point model, the first-harmonic λ-mode eigenvalue separation and the reactivity coupling coefficient were inferred. The present values of these parameters agreed with the results obtained by the reactor noise measurements and the diffusion calculations.

The experimental results show that the magnitudes of the spatial kinetic phenomena including the dependence of the rod reactivity worth on the detector position, the reactivity interaction effect between control rods and the transient flux tilts induced by the rod drop, which have been significantly observed in large thermal and fast power reactors, are inversely proportional to the eigenvalue separation. Applying the two-mode model, the inherent reactivity worths of control rods were also inferred from the space-dependent ones.     

Seismic Analysis of Hydraulic Control Rod Driving System

A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency.     

Transient Analysis for Design of Primary Coolant Pump Adopted to JAERI Passive Safety Reactor JPSR

In a course of a design study of the JAERI passive safety pressurized water reactor (JPSR), a complete loss-of-flow transient caused by a trip of all pumps was analyzed with the RETRAN code to determine an inertia of canned-motor pump utilized as the primary coolant pump and to confirm feasibility of the design condition. This transient was selected because the pump had a low inertia rotor inducing fast flow coastdown, and among the transients in which the pump had dominant effect on the departure from nucleate boiling (DNB), the analyzed transient was severest in view of the DNB occurrence. The DNB threshold was related, based on sensitivity calculations, with the coolant density reactivity coefficient and the pump inertia. From the calculations, it was concluded that the pump inertia higher than 250 kg·m² (8% of the ordinary PWRs) was necessary for preventing the DNB occurrence for the present design of JPSR, regardless of the actuation of the reactor scram. The DNB occurrence could be prevented only by the inherent nature of the reactor core which reduced the power by insertion of negative coolant density reactivity during the transient and this was one of major features of JPSR. It was shown by a rough estimation that the necessary condition could be practically realized by incorporation of a cylindrical-type flywheel.     

Analysis of System Thermal Hydraulic Responses for Passive Safety Injection Experiment at ROSA-IV/Large Scale Test Facility

An experiment was conducted at the ROSA-IV/Large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. The experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top.     

Three-Dimensional Transport Correction in Fast Reactor Core Analysis

Three-dimensional (3-D) transport corrections for K_eff, control rod worths, neutron spectra and power distributions are calculated for the fast critical assemblies ZPPR-9, -10 and -13. The calculations are performed in 7 energy groups using the 3-D and 2-D transport codes TRITAC and TWOTRAN-II. The evaluated 3-D transport corrections are compared with those predicted by a 2-D RZ model and a combined XY, RZ and R model to assess the accuracy of these 2-D models. The RZ model yielded some errors in estimating the transport corrections to K_eff, control rod worths, neutron spectrum and power distribution, while the combined model predicted them with good accuracy.     

Model extension and improvement for simulator-based software safety analysis

One of the major concerns when employing digital I&C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I&C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes “simulator/plant model testing” technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I&C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran–ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then futher be applied to hazard analysis for operator/digital I&C interface interaction failure study, and the hardware-in-the-loop fault injection study.     

System-level hazard analysis using the sequence-tree method

A system-level PHA using the sequence-tree method is presented to perform safety-related digital I&C system SSA. The conventional PHA involves brainstorming among experts on various portions of the system to identify hazards through discussions. However, since the conventional PHA is not a systematic technique, the analysis results depend strongly on the experts’ subjective opinions. The quality of analysis cannot be appropriately controlled. Therefore, this study presents a system-level sequence tree based PHA, which can clarify the relationship among the major digital I&C systems. This sequence-tree-based technique has two major phases. The first phase adopts a table to analyze each event in SAR Chapter 15 for a specific safety-related I&C system, such as RPS. The second phase adopts a sequence tree to recognize the I&C systems involved in the event, the working of the safety-related systems and how the backup systems can be activated to mitigate the consequence if the primary safety systems fail. The defense-in-depth echelons, namely the Control echelon, Reactor trip echelon, ESFAS echelon and Monitoring and indicator echelon, are arranged to build the sequence-tree structure. All the related I&C systems, including the digital systems and the analog back-up systems, are allocated in their specific echelons. This system-centric sequence-tree analysis not only systematically identifies preliminary hazards, but also vulnerabilities in a nuclear power plant. Hence, an effective simplified D3 evaluation can also be conducted.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.